1. Field
This invention pertains generally to nuclear reactor fuel assemblies and more particularly to nuclear reactor fuel assemblies that employ a spacer or mixer or support grid constructed of a high temperature strength, corrosion resistant, accident tolerant composition, and methods of making the spacer or mixer or support grid.
2. Description of Related Art
In most pressurized water nuclear reactors (PWRs), boiling water reactors (BWRs) and heavy water reactors (HWRs), collectively referred to herein as water reactors, the reactor core is comprised of a large number of elongated fuel assemblies that generate the reactive power of the reactor. These fuel assemblies typically include a plurality of fuel rods held in an organized, array by a plurality of grids spaced axially along the fuel assembly length and attached to a plurality of elongated thimble tubes or other support structure of the fuel assembly.
A description of a PWR structure is particularly provided, however, it is understood that the invention is applicable to water reactors in general.
The thimble tubes typically receive control rods or instrumentation therein. Top and bottom nozzles are on opposite ends of the fuel assembly and are secured to the ends of the thimble tubes that extend slightly above and below the ends of the fuel rods.
The grids, as is known in the relevant art, are used to precisely maintain the spacing and support between the fuel rods in the reactor core, provide lateral support for the fuel rods, and induce mixing of the coolant. One type of conventional grid design includes a plurality of interleaved straps that together form an egg-crate configuration having a plurality of roughly square cells which individually accept the fuel rods therein. Depending upon the configuration of the thimble tubes, the thimble tubes can either be received in cells that are sized the same as those that receive fuel rods therein, or in relatively larger thimble cells defined in the interleaved straps. The interleaved straps provide attachment points to the thimble tubes, thus enabling positioning of the grids at spaced locations along the length of the fuel assembly.
The straps are configured such that the cells through which the fuel rods pass each include one or more relatively compliant springs and a plurality of relatively rigid dimples which cooperate to form the fuel rod support feature of the grid. Outer straps of the grid are attached together and peripherally enclose the inner straps of the grid to impart strength and rigidity to the grid and define individual fuel rod cells around the perimeter of the grid. The inner straps are typically welded or braised at each intersection and the inner straps are also welded or braised to the peripheral or outer straps defining the outer perimeter of the assembly.
At the individual cell level, the fuel rods support is normally provided by the combination of rigid support dimples and flexible springs as mentioned above. There are many variations to the spring-dimple support geometry that have been used or are currently in use, including diagonal springs, “I” shaped springs, cantilevered springs, horizontal and vertical dimples, etc. The number of springs per cell also varies. The typical arrangement is two springs and four dimples per cell. The geometry of the dimples and springs needs to be carefully determined to provide adequate rod support through the life of the assembly.
During irradiation, the initial spring force relaxes more or less rapidly, depending on the spring material and irradiation environment. The cladding diameter also changes as a result of the very high coolant pressure and operating temperatures and the fuel pellets inside the rod also change their diameter by densification and swelling. The outside cladding diameter also increases, due to the formation of an oxide layer. As a result of these dimensional and material property changes, maintaining adequate rod support through the life of the fuel assembly is very challenging.
Under the effect of axial flow and cross flow induced by thermal and pressure gradients within the reactor and other flow disturbances, such as standing waves and eddies, the fuel rods, which are slender bodies, are continuously vibrating with relatively small amplitudes. If the rod is not properly supported, this very small vibration amplitude may lead to relative motion between the support points and the cladding. If the pressure exerted by the sliding rod on the relatively small dimple and grid support surfaces is high enough, the small corrosion layer on the surface of the cladding can be removed by abrasion, exposing the base metal to the coolant. As a new corrosion layer is formed on the exposed fresh cladding surface, it is also removed by abrasion until ultimately the wall of the rod is perforated. This phenomenon is known as corrosion fretting and in 2006 it was the leading cause of fuel failures in PWR reactors.
Support grids also provide another important function in the fuel assembly, that of coolant mixing to decrease the maximum coolant temperature. Since the heat generated by each fuel rod is not uniform, there are thermal gradients in the coolant. One important parameter in the design of the fuel assemblies is to maintain the efficient heat transfer from the fuel rods to the coolant. The higher the amount of heat removed per unit time, the higher the power being generated. At high enough coolant temperatures, the rate of heat that can be removed per unit of cladding area in a given time decreases abruptly in a significant way. This phenomenon is known as deviation from nucleate boiling or DNB. If within the parameters of reactor operation, the coolant temperature would reach the point of DNB, the cladding surface temperature would increase rapidly in order to evacuate the heat generated inside the fuel rod and rapid cladding oxidation would lead to cladding failure. It is clear that DNB needs to be avoided to prevent fuel rod failures. Since DNB, if it occurs, takes place at the point where the coolant is at its maximum temperature, it follows that decreasing the maximum coolant temperature by coolant mixing within the assembly permits the generation of larger amounts of power without reaching DNB conditions. Normally, the improved mixing is achieved by using mixing vanes in the down flow side of the grid structure. The effectiveness of mixing is dependent upon the shape, size and location of the mixing vanes relative to the fuel rod.
Other important functions of the grid include the ability to sustain handling and normal operation at anticipated accident loads without losing function and to avoid “hot spots” on the fuel rods due to the formation of steam pockets between the fuel rods and the support points, which may result when not enough coolant is locally available to evacuate the heat generated in the fuel rod. Steam pockets cause overheating of the fuel rod to the point of failure by rapid localized corrosion of the cladding.
The grids, grid straps and integral flow mixers, e.g., mixing vanes, typically have been constructed of zirconium alloy because these materials exhibit low neutron adsorption cross-section and adequate mechanical and chemical properties. Similarly, fuel cladding materials also have been constructed of zirconium alloy. However, alternative fuel cladding materials are being considered for future nuclear reactor design and operation. Such new and different materials include silicon carbide (SiC) ceramic matrix composites which demonstrate properties that can provide for better safety margin and accident tolerance. However, the benefits of implementing new fuel cladding materials, such as SiC, can be offset because the grids, straps and/or mixing vanes inside the core contain a significant amount of zirconium. Thus, it is desirable to replace the zirconium-containing grids, straps and mixing vanes with other materials which have better structure stability, strength, and oxidation resistance at temperatures beyond normal operation and design basis accidents of a nuclear reactor.
It is thus desired to provide an improved material (e.g., containing little to no zirconium) that exhibits high temperature strength, corrosion resistance and accident tolerance for use in constructing grids for nuclear reactor fuel assemblies.